GO2-13-151, NEDO-33813, Rev. 2, Technical Specification Change Support for Rhr/Lpci and LPCS Flow Rate. Long-Term LOCA Containment Response and ECCS/Non-LOCA Evaluations.

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NEDO-33813, Rev. 2, Technical Specification Change Support for Rhr/Lpci and LPCS Flow Rate. Long-Term LOCA Containment Response and ECCS/Non-LOCA Evaluations.
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LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORE COOLING SYSTEMS. SUO".LLANCE REQUIREMENTS, GE Hitachi Nuclear Energy NEDO-33813 Revision 2 September 2013 Non-Propriety Information - Class I (Public)

Technical Specification Change Support for RHR/LPCI and LPCS Flow Rate Long-Term LOCA Containment Response and ECCS/Non-LOCA Evaluations

GE Hitachi Nuclear Energy 0 HITACHI NEDO-33813 Revision 2 eDRF Section 0000-0156-3022 R6 September 2013 Non-ProprietaryInformation - Class I (Public)

Technical Specification Change Support for RIIR/LPCI and LPCS Flow Rate Long-Term LOCA Containment Response and ECCS/Non-LOCA Evaluations Copyright 2013 GE-HitachiNuclearEnergy Americas LLC All Rights Reserved

NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public)

INFORMATION NOTICE This is a non-proprietary version of the document NEDC-33813P, Revision 2, which has the proprietary information removed. Portions of the document that have been removed are indicated by an open and closed bracket as shown here (( fl.

IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The design, engineering, and other information contained in this document is furnished for the purposes of supporting the Energy Northwest pursuit of a license amendment to change the flow rate for the RHR/LPCI and LPCS in proceedings before the U. S. Nuclear Regulatory Commission. The only undertakings of GEH with respect to information in this document are contained in the contract between GEH and Energy Northwest, and nothing contained in this document shall be construed as changing that contract. The use of this information by anyone for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use, GEH makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document.

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NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public)

REVISION

SUMMARY

Rev. Required Changes to Achieve Revision 0 NA Incorporate responses to customer comments on Revision 0 to add clarification and minor corrections.

2 Corrected typographical errors in Tables 1 and 2.

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Table of Contents Section Page 1.0 Scope and Summ ary ........................................................................................................ 1 1.1 Project Summary .................................................................................................... 1 1.2 Task Scope .................................................................................................................. 1 1.3 Results Summary .................................................................................................... 2 1.3.1 Key Evaluation Results ...................................................................................... 2 1.3.2 Key Results Outside Safety or Design Lim its ...................................................... 2 1.3.3 Direct Effect on Plant Configuration ................................................................... 2 1.3.4 Effect on Design Operating Margins .................................................................... 3 1.3.5 Effect on Other Tasks ........................................................................................... 3 1.3.6 Implem entation Recomm endations ...................................................................... 3 1.3.7 Observations ........................................................................................................ 3 1.3.8 Effect on Equipment Out-of-Service and Performance Improvement Features ......... 3 2.0 Long-Term DBA LOCA Containment Response Analyses ......................................... 4 2.1 Analyses Description and Methodology ................................................................. 4 2.2 Input and Assumptions ........................................................................................... 5 2.2.1 Key Input ......................................................................................................... 5 2.2.2 Key Assumptions ............................................................................................... 7 2.3 Results ........................................................................................................................ 8 2.3.1 Key Results ........................................................................................................ 8 2.3.2 Supporting Evaluations ..................................................................................... 27 2.4 Recommendations and Observations 27 27................................

2.4.1 Recommendations ............................................................................................. 27 2.4.2 Observations ...................................................................................................... 27 3.0 ECCS-LOCA Analyses Assessment ........................................................................ 28 4.0 Non-LOCA Assessments .......................................................................................... 29 4.1 M ethodology and Disposition ................................................................................ 29 4.1.1 Environmental Qualification/High Energy Line Break Assessment ................... 29 4.1.2 AOO Assessment ............................................................................................... 29 4.1.3 ATW S Assessment ........................................................................................... 30 iv

NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public)

Table of Contents Section Page 4.1.4 Suppression Pool Local Heating (NUREG-0783) Assessment .......................... 30 4.1.5 Shutdown Cooling Mode Assessment ............................................................... 31 4.1.6 Alternate Shutdown Cooling Mode Assessment ............................................... 32 4.1.7 Refueling Operations and Spent Fuel Pool Cooling Assessment ........................ 33 4.1.8 Appendix R/Safe Shutdown Fire Assessment ................................................... 33 4.1.9 Nuclear Boiler System Assessment .................................................................... 33 4.2 Recommendations and Observations ...................................................................... 34 4.2.1 Recommendations ............................................................................................. 34 4.2.2 Observations ..................................................................................................... 34 5.0 Conclusions .................................................................................................................. 35 6.0 References .................................................................................................................... 36 V

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List of Tables Table Page Table 1. Key Inputs for the Long-Term Containment Response ............................................. 5 Table 2. A ssum ptions ......................................................................................................... 7 Table 3. Long-Term Analysis Results .................................................................................... 9 List of Figures Figure Page Figure 1. Case A Containment Pressure Responses and Differential Pressure ....................... 10 Figure 2. Case A DW Temperature Response ........................................................................... 11 Figure 3. Case A SP Temperature Response ........................................................................ 12 Figure 4. Case A RH R Rate ................................................................................................. 13 Figure 5. Case B Containment Pressure Responses and Differential Pressure ........................ 14 Figure 6. Case B DW Temperature Response ........................................................................ 15 Figure 7. Case B SP Temperature Response ........................................................................ 16 Figure 8. Case B RH R Rate ................................................................................................. 17 Figure 9. Case C Containment Pressure Responses and Differential Pressure ......................... 18 Figure 10. Case C DW Temperature Response ..................................................................... 19 Figure 11. Case C SP Temperature Response ...................................................................... 20 Figure 12. Case C RH R Rate ............................................................................................... 21 Figure 13. Case C Long-Term Maximum Pressure in the DW and RPV (30 Days) ............... 22 Figure 14. Case D Containment Pressure Responses and Differential Pressure .................... 23 Figure 15. Case D DW Temperature Response ................................................................... 24 Figure 16. Case D SP Temperature Response ..................................................................... 25 Figure 17. C ase D RH R Rate ............................................................................................... 26 vi

NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public)

Acronyms and Abbreviations Short Form Description ANS American Nuclear Society AOO Anticipated Operational Occurrence ARTS/MELLLA Average Power Range Monitor/Rod Block Monitor/Technical Specifications/Maximum Extended Load Line Limit Analysis ATWS Anticipated Transient Without Scram CGS Columbia Generating Station DBA Design Basis Accident DW Drywell ECCS Emergency Core Cooling System EQ Environmental Qualification FSAR Final Safety Analysis Report FW Feedwater GEH GE Hitachi Nuclear Energy gpm Gallon per minute HELB High Energy Line Break hp Horsepower HPCS High Pressure Core Spray LOCA Loss-of-Coolant Accident LPCI Low Pressure Coolant Injection LPCS Low Pressure Core Spray LWL Low Water Level MELLLA Maximum Extended Load Line Limit Analysis MSIV Main Steam Isolation Valve MWt Megawatt Thermal vii

NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public)

Acronyms and Abbreviations Short Form Description NRC Nuclear Regulatory Commission PRNM Power Range Neutron Monitor psia Pounds per square inch absolute psid Pounds per square inch difference psig Pounds per square inch gage RCS Reactor Coolant System RHR Residual Heat Removal RPV Reactor Pressure Vessel RSLB Recirculation Suction Line Break SC Safety Communication SIL Services Information Letter SP Suppression Pool SR Surveillance Requirement SRV Safety Relief Valve SSC System, Structure, or Component TS Technical Specification WW Wetwell OF Degree Fahrenheit Viii

NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public) 1.0 Scope and Summary 1.1 Project Summary The GE Hitachi Nuclear Energy (GEH) services described in this document support Energy Northwest's pursuit of a license change to lower the flow rate Technical Specification (TS) embodied in surveillance requirement (SR) 3.5.1.4 and SR 3.5.2.5 for the low pressure coolant injection (LPCI) mode of residual heat removal (RHR) resulting in more operational margin.

Current TS under SR 3.5.1.4 and SR 3.5.2.5 for LPCI are > 7,450 gpm with 20 psid between the reactor and drywell (DW) (26 psid between the reactor pressure vessel (RPV) and suppression pool (SP)). The proposed TS value is 7,200 gpm, supported by an analytical value of 6,713 gpm, with 20 psid between the reactor and DW (26 psid between the RPV and SP).

The GEH scope of work for the support of a TS change includes long-term containment analyses and assessments for all other RHR modes of operation affected by the proposed TS change. An emergency core cooling system (ECCS) - loss-of-coolant accident (LOCA) performance evaluation with a reduced LPCI flow rate and a reduced low pressure core spray (LPCS) flow rate is considered.

1.2 Scope The purpose of this evaluation is to perform long-term LOCA containment analyses for Columbia Generating Station (CGS) for Final Safety Analysis Report (FSAR) Cases A, B and C at the new RHR/LPCI analytical flow rate value of 6,713 gpm with 20 psid between the reactor and DW (26 psid between the RPV and SP). (Reference 1). An additional case has been performed to address Safety Communication (SC) 06-01 (Reference 2). This case is referred to as "Case D." The intent of Cases A, B, and C is to document, at the analytical flow rate of 6,713 gpm (which is bounded by the proposed 7,200 gpm TS value), that the current FSAR (Reference

1) peak SP temperature (long-term) of 204.5°F is not increased.

An ECCS-LOCA fuel analysis assessment is performed to confirm that an analytical LPCI flow rate of 6,713 gpm with 20 psid between the reactor and DW (26 psid between the RPV and SP) is acceptable. Additionally, confirm that the current TS LPCS flow rate of > 6,350 gpm with 122 psid between the reactor and DW (128 psid between the RPV and SP) per TS 3.5.1 and 3.5.2 can be reduced to an analytical flow rate of 5,625 gpm with 122 psid between the reactor and DW (128 psid between reactor and SP), bounding the proposed TS value of> 6,200 gpm with 122 psid between the reactor and DW (128 psid between the RPV and SP).

Non-LOCA analyses that may be affected by the TS flow rate change have been assessed. The assessments include the following analyses:

  • Environmental Qualification (EQ) /High Energy Line Break (HELB)

" Anticipated Operational Occurrences (AOOs)

" SP Local Heating (NUREG-0783)

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NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public)

" Shutdown Cooling Mode

" Alternate Shutdown Cooling Mode

" Refueling Operations and Spent Fuel Pool Cooling

  • Appendix R/Safe Shutdown Fire

" Nuclear Boiler System The effects on the following plant changes and flexibility options have been evaluated:

" PRNM and ARTS/MELLLA

" Operating Flexibility Options

" Elimination of Selected Response Time Requirements (NEDO-32291)

" SLCS

" MSIV Closure on Level 1 1.3 Results Summary 1.3.1 Key Evaluation Results The following results were demonstrated to remain within the Safety and Design Limits

  • Long-Term Design Basis Accident (DBA) LOCA Response o SP Temperature
  • ECCS-LOCA Fuel Analysis o Existing analysis results remain valid

" Non-LOCA Analyses o Existing analysis results remain valid or have an insignificant effect based on assessment results.

1.3.2 Key Results Outside Safety or Design Limits None 1.3.3 Direct Effect on Plant Configuration The results of this evaluation demonstrate that a change in the RHR/LPCI and LPCS flow rates to analytical values of 6,713 gpm and 5,625 gpm respectively is satisfactory. These analytical flow rates bound the proposed TS RHR/LPCI and LPCS flow rates of 7,200 gpm and 6,200 gpm respectively.

The GE RHR System Design Specification for Columbia remains applicable as contributing to the Columbia RHR system TS value calculations, assuming no other changes were made in the associated plant configuration.

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NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public) 1.3.4 Effect on Design Operating Margins The results of this evaluation confirm that a RHR/LPCI and LPCS flow rate TS change has an insignificant or no reduction on operating margin with respect to SP temperature.

1.3.5 Effect on Other Tasks Any indirect effects on other systems, structures, or components (SSCs) and design features are outside the scope of this report.

1.3.6 Implementation Recommendations None 1.3.7 Observations None 1.3.8 Effect on Equipment Out-of-Service and Performance Improvement Features None 3

NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public) 2.0 Long-Term DBA LOCA Containment Response Analyses 2.1 Analyses Description and Methodology The long-term containment response for FSAR Cases A, B and C are re-analyzed with a LPCI flow rate of 6,713 gpm. The purpose of this analysis is to show that with a LPCI flow rate of 6,713 gpm and other updated inputs do not cause the SP temperature to exceed the current FSAR (Reference 1) value of 204.5°F.

An additional case to assess the effect of Safety Communication SC 06-01 (Reference 2) on the SP design margin is also evaluated. This case is called "Case D" and is used to determine how long CGS can run all ECCS pumps with one RHR heat exchanger available before turning off the additional pumps to stay within the results of FSAR (Reference 1) Case C.

A description of each case is provided below:

Case A: All ECCS equipment operating - with containment spray This case assumes that offsite power is available to operate all cooling systems. During the first 600 seconds following the pipe break, the high pressure core spray (HPCS), LPCS, and all LPCI pumps are assumed operating. All flow is injected directly into the RPV. After 600 seconds, both RHR heat exchangers are activated to remove energy from containment. During this mode of operation the flow from two of the LPCI pumps is routed through the RHR heat exchangers where it is cooled before it is discharged into the containment spray header. HPCS is always available.

Case B: Loss of offsite power - with delayed containment spray This case assumes no offsite power is available following the accident and that only the HPCS and LPCI diesels (Divisions 3 and 2, respectively) are available. For the first 600 seconds following the break the HPCS and two LPCI pumps are used exclusively for core cooling.

During this time an operational restriction prohibits the activation of containment spray. After 600 seconds, one RHR heat exchanger is activated. The flow from one pump is routed through the heat exchanger and is discharged into the containment spray line. The second LPCI pump is assumed to be shut down. HPCS is always available.

Case C: Loss of offsite power - no containment spray This case assumes no offsite power is available following the accident and that only the HPCS and LPCI diesels are available. For the first 600 seconds following the accident, one HPCS and two LPCI pumps are used exclusively to cool the core. After 600 seconds, one RHR heat exchanger is activated to remove energy from the containment, but containment spray is not activated. The LPCI flow cooled by the RHR heat exchanger is discharged into the RPV. The second LPCI pump is assumed to be shut down. HPCS is always available.

Case D: All ECCS equipment operating - RHR heat exchanger failure

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The GEH computer code SHEX is used to analyze the long-term DBA LOCA containment response. References 3 and 4 document the SHEX application methodology. Reference 5 documents the Nuclear Regulatory Commission (NRC) acceptance of the application of SHEX for containment analyses.

2.2 Input and Assumptions 2.2.1 Key Input Table 1 provides a brief summary of the key inputs used in the long-term containment analysis.

Table 1. Key Inputs for the Long-Term Containment Response No Parameter Unit Analysis Value

1. Reactor
a. Initial power level 102% current rated power MWt 3,556
b. Initial vessel dome pressure psia 1,055
c. Decay heat model American Nuclear Society (ANS) 5.1-1979 +

2 sigma and Services Information Letter (SIL) 636 (Reference 6)

d. Vessel volumes
1. Total vessel free volume ft3 23,679
2. Liquid vessel volume Subcooled ft3 8138 Saturated ft3 3941
e. LOCA break area fi2 3.1893
f. Time at which main steam isolation valves (MSIVs) start to sec close / Fully closed 0.0/3.0 Recirculation Suction Line Break (RSLB) 5

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No Parameter Unit Analysis Value

2. DW/Vent System
a. Total DW free volume (including vent system) Wt3 200,540
b. Initial DW pressure psig 2.00
c. Initial DW temperature OF 150
d. Initial DW relative humidity  % 20
e. Vent System
1. Number of downcomers 99
2. Inside diameter of each downcomer ft 1.995
3. Loss coefficient for vent system (including entrance 2.770 and exit losses)
3. Wetwell (WW)/SP
a. Initial SP volume (including water in vents)
1. Low water level (LWL) ft3 107,850
b. Initial SP temperature (max) OF 90
c. Initial WW airspace volume
1. LWL Wf 3 144,184
d. Initial WW/containment airspace pressure psig 2.00
e. Initial WW/containment airspace temp OF 90
f. Initial WW/containment airspace relative humidity  % 100
4. LPCI/LPCS/HPCS Pump Heat
a. LPCI pump horsepower hp 850
b. LPCS pump horsepower hp 1,500
c. HPCS pump horsepower hp 3,000
5. LPCI/LPCS/HPCS Flow Rate Per Loop
a. LPCI or DW/WW Spray Flow Rate gpm 6,713'
b. LPCS Flow Rate gpm 5,6252
c. [HPCS Flow Rate gpm 6,250' 6

NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public)

No Parameter Unit Analysis Value

6. RHR
a. Service water temperature OF 85 for first 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> of I_ LOCA, 90 thereafter.
b. Heat exchanger K-Value for LOCA with service water Btu/sec- SP K temperature of 85°F OF 125 284.5 SP Temperature 'F versus K-Value (K) 165 287 185 288.1 200 288.8 I.Flow rate at 20 psid between the reactor and DW (26 psid between the RPV and SP) 2.Flow rate at 122 psid between the reactor and DW (128 psid between reactor and SP) 3.Flow rate at 194 psid between the reactor and DW (200 psid between reactor and SP) 2.2.2 Key Assumptions Table 2 provides a brief summary of the key assumptions of the long-term containment analysis.

Table 2. Assumptions Item Parameter Reference/Basis 1 The reactor is operating at 102% of the current licensing power See Table 1, (3,556 MWt). Item L.a 2 The LOCA is initiated by an instantaneous double-ended Current license basis rupture of the recirculation suction line.

3 The vessel break flow rates during the blowdown are based on Current license basis the Moody slip critical flow model (Reference 7) 4 Core decay heat is based on ANS 5.1-1979 + 2 sigma.

Additional decay heat terms per GEH SIL 636 are included in Reference 6 the decay heat.

5 MSIV closure starts at 0.0 seconds after the start of the See Table 1, accident and full closure is achieved at 3.0 seconds following Item 1.f closure.

6 Feedwater (FW) flow into the vessel continues until all the Current license basis high energy FW (above 200'F) is injected into the vessel.

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Item Parameter Reference/Basis 7 Thermodynamic equilibrium exists between liquids and gases in the DW and between the SP and the suppression chamber Current license basis airspace.

8 After downcomer clearing, downcomer flow to the SP consists Current license basis of a homogeneous mixture of the fluid in the DW.

9 The initial SP volume is at the minimum TS limit value to Current license basis maximize the long-term SP temperature response.

10 The initial SP temperature is at the maximum TS limit value. Current license basis 11 The ECCS operation assumptions are described in Section 2.1. Current license basis 12 No credit is taken for passive heat sinks in the DW, the SP, or --

the suppression chamber airspace.

2.3 Results Table 3 provides a brief summary of the key results of the long-term containment analysis.

2.3.1 Key Results Similar to the FSAR analysis in Reference 1, the most limiting case with respect to SP temperature is Case C. Case C yields a maximum peak pool temperature of (( )),

which is below the maximum pool temperature reported in the CGS FSAR (Reference 1). The results for the other key parameters remain below their respective design limits.

The results for Cases A, B, C and D are shown in Figures 1 through 4, Figures 5 through 8, Figures 9 through 13, and Figures 14 through 17, respectively.

Figure 13 shows the long-term maximum DW and RPV pressure for the limiting case (Case C) for 30 days. In the long-term the DW and RPV pressures are depicted as being the same. The pressure presented in the figure is the maximum pressure present in the RPV and DW.

In order to address SC 06-01 (Reference 2), ((

)) in a Case D scenario after the onset of the LOCA to keep the SP temperature below the maximum calculated value for FSAR (Reference 1) Case C.

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Table 3. Long-Term Analysis Results Parameters Case A Case B Case C Case D Design Item Limit Note:

(1) This is the FSAR (Reference 1) maximum pool temperature for long-term containment analyses.

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NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public) 2.3.1.1 DBA LOCA Case A Results - Figures 1[

Figure 1. Case A Containment Pressure Responses and Differential Pressure 10

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[1 Figure 2. Case A DW Temperature Response 11

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It 1]

Figure 3. Case A SP Temperature Response 12

NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public) ci Figure 4. Case A RHR Rate 13

NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public) 2.3.1.2 DBA LOCA Case B Results - Figures 11 1]

Figure 5. Case B Containment Pressure Responses and Differential Pressure 14

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Figure 6. Case B DW Temperature Response 11 15

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Figure 7. Case B SP Temperature Response 16

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Figure 8. Case B RHR Rate 17

NEDO-33813 Revision 2 Non-Proprietary Information - Class .1(Public) 2.3.1.3 DBA LOCA Case C Results - Figures 11 Figure 9. Case C Containment Pressure Responses and Differential Pressure 18

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[r Figure 10. Case C DW Temperature Response 19

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Figure 11. Case C SP Temperature Response 20

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Figure 12. Case C RHR Rate 21

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[1 Figure 13. Case C Long-Term Maximum Pressure in the DW and RPV (30 Days) 22

NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public) 2.3.1.4 DBA LOCA Case D Results - Figures s[r Figure 14. Case D Containment Pressure Responses and Differential Pressure 23

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Figure 15. Case D DW Temperature Response 24

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Figure 16. Case D SP Temperature Response 25

NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public) 11 11 Figure 17. Case D RHR Rate 26

NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public) 2.3.2 Supporting Evaluations None 2.4 Recommendations and Observations 2.4.1 Recommendations

(( )) in a Case D scenario after the onset of the LOCA.

2.4.2 Observations None 27

NEDO-33813 Revision 2 Non-Proprietary Information - Class .I (Public) 3.0 ECCS-LOCA Analyses Assessment The ECCS LOCA fuel analysis assessment supporting Reference 8 confirms that a TS LPCI flow rate > 6,713 gpm with 20 psid between the reactor and DW (26 psid between the RPV and SP) is acceptable.

The ECCS LOCA fuel analysis assessment supporting Reference 8 confirms that a TS LPCS flow rate > 5,625 gpm with 122 psid between the reactor and DW (128 psid between the RPV and SP) is acceptable.

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NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public) 4.0 Non-LOCA Assessments 4.1 Methodology and Disposition Non-LOCA events were assessed to determine the effect of the TS RHRILPCI and LPCS flow rate changes as applicable. The referenced licensing basis analysis conditions, before the reduction in RHR/LPCI and LPCS flow rates, are assumed to remain valid for the assessments.

The assessments made in this section are only for the effect of RHR/LPCI and LPCS flow rate changes. The events that are assessed for this report are those that are potentially affected by a RHR/LPCI and LPCS flow rate changes as applicable. The long-term LOCA and ECCS-LOCA analyses are addressed in Sections 2.0 and 3.0. The following assessments were performed for non-LOCA events and the dispositions are provided.

4.1.1 Environmental Qualification/High Energy Line Break Assessment HELBs do not include RHR/LPCI and LPCS lines, therefore the HELBs or EQ outside the containment is not affected. For EQ inside the containment, the long-term analyses results provided in Section 2.0 can be used as inputs for the pressure and temperature envelopes as applicable.

4.1.2 AOO Assessment CGS FSAR includes the following AOO categories:

  • Events resulting in decrease in coolant temperature
  • Events resulting in increase in reactor pressure
  • Reactivity and power distribution anomalies Events resulting in decrease in coolant temperature are:
  • Loss of FW heating
  • FW controller failure
  • Pressure regulator failure
  • Inadvertent RHR shutdown cooling The first four events are relevant to the high pressure operation only. RHR/LPCI and LPCS flow is not credited in the analyses of these events. Inadvertent RHR shutdown cooling is also not adversely affected by a reduction in RHR flow rate, because a reduction in the flow rate makes the event only milder than initially analyzed. Therefore, there is no effect of RHR/LPCI and LPCS flow rate reduction on the events resulting in decrease in coolant temperature.

Events resulting in increase in reactor pressure occur at high pressure and of very short duration (within minutes). The scenarios do not rely on LPCI or LPCS flow. Therefore there is no effect of RHR/LPCI and LPCS flow rate reduction on the events resulting in increase in reactor pressure.

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Events resulting in decrease in RCS flow occur due to a failure in the recirculation loop, and are relevant to operation at high RCS pressure. The analyses do not credit RHR/LPCI and LPCS flow, and therefore not affected by a reduction in RHRILPCI and LPCS flow rate.

Reactivity and power distribution anomalies are relevant to the normal operation at high RCS pressure. The analyses do not credit RHR/LPCI and LPCS flow, and therefore not affected by a reduction in RHR/LPCI and LPCS flow rate.

Based on the above discussion, it can be concluded that AOOs are not affected by the RHR/LPCI and LPCS flow rate reduction, and no further evaluation of AQOs is required.

4.1.3 ATWS Assessment ATWS event analyses credit heat removal by the RHR heat exchangers for SP cooling. The peak SP temperatures for ATWS are calculated ((

The ATWS analysis presented in the FSAR (Reference 9) uses a K-value of 578 Btu/sec-°F for two heat exchangers with a service water temperature of 90'F. An evaluation using a reduced heat exchanger K-value of ((

)) shows that the pool temperature would have an insignificant increase

)). This increase in pool temperature is offset by a reduced service water temperature of 85°F (for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> following the event) and corresponding temperature dependent K-values. The service water temperature and corresponding K-values provide more heat removal capability than the service water temperature and K-value used in the FSAR ATWS analysis.

The net effect is that the peak SP temperature does not exceed the limiting FSAR ATWS analysis value of 173.4°F. Therefore, the FSAR ATWS analysis for peak SP temperature remains valid, and there is no adverse effect on the capability of the plant systems to mitigate postulated ATWS events.

Note that the peak SP temperature for the limiting ATWS case presented in Reference 10 would be affected by a maximum of ((

)). This increase in pool temperature is offset by a reduced service water temperature of 85°F (for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> following the event) and corresponding temperature dependent K-values. The service water temperature and corresponding K-values provide more heat removal capability than the service water temperature and K-value used in the Reference 10 ATWS analysis. The net effect is that the peak SP temperature does not exceed the limiting Reference 10 ATWS analysis value of 187°F. The CGS operation in the maximum extended load line limit analysis (MELLLA) region has no adverse effect on the capability of the plant systems to mitigate postulated ATWS events.

4.1.4 Suppression Pool Local Heating (NUREG-0783) Assessment CGS is equipped with SRV X-quenchers; NUREG-0783 analysis is not required according to Reference 11, which is referenced in the FSAR. However, an assessment of the RHR/LPCI flow 30

NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public) rate change on the analysis of record from Reference 11 for SP local heating (NUREG-0783) is provided below.

The SP local heating analysis assumes one RHR heat exchanger is available for pool cooling.

The RHR heat exchanger K-value can be affected by a change in the RHR flow rate.

The SP local heating analysis from Reference 11 uses a K-value of 289 Btu/sec-°F for the heat exchanger with a service water temperature of 90°F. The updated RHR flow rate (6,713 gpm) corresponds to a heat exchanger K-value that is a maximum of 4.5 Btu/sec-0 F less than the Reference 11 K-value for pool cooling. As shown in Section 4.1.3, small changes in K-value have an insignificant effect on SP temperature. An increase in SP temperature due to the difference in K-values above is offset by the updated service water temperature for pool cooling (850 F for ten hours following the event). The service water temperature and corresponding K-values provide more heat removal capability than the service water temperature and corresponding K-value used in the Reference 11 SP local heating analysis.

Based on the above assessment, the net effect is that the peak SP temperature does not exceed the Reference II SP local heating (NUREG-0783) analysis value of (( )). Therefore, the Reference 11 SP local heating (NUREG-0783) analysis is not adversely affected by a RHR/LPCI flow rate change.

4.1.5 Shutdown Cooling Mode Assessment The shutdown cooling mode can be affected by a change in the RHR flow rate if there is a corresponding reduction in the RHR heat exchanger K-value. A change in the heat exchanger K-value may increase the time required for the RHR system to cool the RPV down to 125°F.

The RHR heat exchanger K-value (289 Btu/sec-°F per heat exchanger) used in the power uprate analysis for shutdown cooling (Reference 11, Section 21.4.1.3) is reduced by a maximum of 5.4 Btu/sec-°F for shutdown cooling mode due to the reduction in RHR flow rate.

The Reference 11 analysis time to cool the RPV down to 125'F is (( )). Using a reduced heat exchanger K-value (284 Btu/sec-°F), ((

)). The aforementioned K-value and service water temperature combination is conservative compared to the current K-value and service water temperature combination with respect to reactor cool down time. The service water temperature for shutdown cooling is 80'F with a heat exchanger K-value of 283.6 Btu/sec-°F. The 80'F service water temperature offsets the increase in time to reach 125 0 F due to the reduced K-value. The service water temperature and corresponding K-value provide more heat removal capability than the service water temperature and corresponding K-value used in the Reference 11 shutdown cooling analysis. The net effect is that the reactor cool down time does not exceed the Reference 11 analysis value of ((

Therefore, the Reference 11 analysis for reactor cool down time remains valid and the RHR system remains capable of performing its intended function with the reduced RHR/LPCI flow rate.

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NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public) 4.1.6 Alternate Shutdown Cooling Mode Assessment The alternate shutdown cooling mode can be affected by a change in the RHR flow rate if there is a corresponding reduction in the RHR heat exchanger K-value. A change in the heat exchanger K-value may affect the peak bulk SP temperature and the time to cool the RPV to cold shutdown.

The RHR heat exchanger K-value (289 Btu/sec-°F per heat exchanger) used in the power uprate analysis for alternate shutdown cooling (Reference 11, Section 21.4.1.3) is reduced by a maximum of 4.5 Btu/sec-0 F due to the reduction in RHR flow rate. The K-value and service water temperature are different than the values for normal shutdown cooling.

The power uprate alternate shutdown cooling mode analysis (Reference 11) demonstrates that the time required for the RHR system to cool the RPV to cold shutdown is (( )) for Activity Cl -b2 compared to the Regulatory Guide 1.139 requirement of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. For the alternate shutdown cooling event at power uprate operating conditions, the calculated peak SP bulk temperature is (( )) for Activity C2/CI which is lower than the FSAR value of

(( )) for the original licensed thermal power reported in Reference 11, Table 29-5. This difference in temperature provides margin between the FSAR value reported in Reference 11 Table 29-5 and the power uprate value. The FSAR value remains bounding and does not need to be changed.

The power uprate Activity Cl-b2 cold shutdown time is not adversely affected by the reduced heat exchanger K-value because it is offset by the updated service water temperature for pool cooling (85°F for ten hours following the event. ((

))). The service water temperature and corresponding K-values provide more heat removal capability than the service water temperature and corresponding K-value used in the Reference 11 alternate shutdown cooling analysis (Activity Cl-b2). The net effect is that the cold shutdown time does not exceed the Reference II analysis value of(( )). The cold shutdown time remains below the Regulatory Guide 1.139 requirement and the current FSAR cold shutdown time of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The power uprate analysis Activity C2/C1 shows a )) to the peak bulk SP temperature FSAR value for the original licensed thermal power presented in Reference 11, Table 29-5. A decrease in the heat exchanger K-value would have a maximum effect on the peak SP temperature of (( )). An increase in SP temperature due to the difference in K-values is offset by the updated service water temperature of 85°F for ten hours following the event. The service water temperature and corresponding K-values provide more heat removal capability than the service water temperature and corresponding K-value used in the Reference 11 alternate shutdown cooling analysis (Activity C2/C1). The net effect is that the peak SP temperature does not exceed the Reference 11 analysis value of (( )) and the time to cold shutdown remains below 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> for Activity C2/Cl. None of the available margin is used.

Therefore the alternate shutdown cooling mode is able to perform its intended function and the FSAR analysis remains valid and is not adversely affected by a reduced RHR/LPCI flow rate.

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NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public) 4.1.7 Refueling Operations and Spent Fuel Pool Cooling Assessment Refueling operations and spent fuel pool cooling may be affected if there is a reduction in the RHR heat exchanger heat removal capacity as a result of the RHR/LPCI and LPCS flow rate reduction. Because GEH does not have the plant records for refueling operations, these evaluations are not included in this report. These assessments are made by Energy Northwest outside of this report.

4.1.8 Appendix R/Safe Shutdown Fire Assessment The Appendix R / safe shutdown fire analysis can be affected by a change in the RHR flow rate and a corresponding reduction in the RHR heat exchanger K-value. A change in the RHR flow rate and heat exchanger K-value may affect the peak cladding temperature and peak bulk SP temperature.

The RHR flow rate used in the current Appendix R analysis for peak cladding temperature is the same as the updated RHR flow rate. Therefore the Appendix R analysis for peak cladding temperature remains valid.

The heat exchanger K-value (289 Btu/sec-0 F) used in the current Appendix R analysis for SP temperature is reduced by a maximum of 4.5 Btu/sec-0 F. The K-value and service water temperature are different than the values for normal shutdown cooling.

The reduction in K-value would result in a maximum increase in the SP temperature of ((

The increase would not exceed the SP design limit of 212'F. An increase in SP temperature due to the difference in K-values is offset by the updated service water temperature of 850 F for ten hours following the event. The service water temperature and corresponding K-values provide more heat removal capability than the service water temperature and corresponding K-value used in the Appendix R analysis. The net effect is that the peak SP temperature does not exceed the current Appendix R analysis value of(( )). There is a lot of margin between the peak SP temperature of (( )) and the SP design limit for Appendix R (212 0 F).

Therefore the Appendix R analysis remains valid and the reduced RHR/LPCI flow rate does not adversely affect the ability of the ECCS to mitigate this event and safe shutdown of the reactor can be achieved.

4.1.9 Nuclear Boiler System Assessment This section assesses the effect of a RHR/LPCI and LPCS flow rate change on the CGS nuclear boiler system. The scope of this assessment is limited to the nuclear boiler instruments, MSIVs, main steam flow restrictors, and safety relief valves. This is consistent with the assessment provided in Section 8.0 of Reference 11.

MSIVs, main steam flow restrictors, and safety relief valves are relevant at high RCS pressure and therefore not affected by a reduction in RHR/LPCI and LPCS flow rate. Nuclear Boiler Instruments operation is not affected by RHR/LPCI and LPCS flow rate.

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NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public) 4.2 Recommendations and Observations 4.2.1 Recommendations None 4.2.2 Observations None 34

NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public) 5.0 Conclusions The reanalyzed long-term LOCA containment cases for CGS FSAR, Cases A, B and C, support a RHR/LPCI analytical flow rate value of 6,713 gpm with 20 psid between the reactor and DW (26 psid between the RPV and SP). Cases A/B/C document that an analytical flow rate of 6,713 gpm, bounding the proposed 7,200 gpm TS value, does not cause the current FSAR (Reference 1) peak SP temperature (long-term) of 204.5°F to be increased.

The additional Case D that is performed to address SC 06-01 (Reference 2) indicates ((

)) in a Case D scenario after the onset of the LOCA to keep the SP temperature below the maximum calculated value for FSAR Case C.

The ECCS-LOCA fuel analysis assessment confirms that the proposed TS LPCI flow rate of 7,200 gpm, supported by an analytical value of 6,713 gpm is acceptable. Additionally, it confirms that the current TS LPCS flow rate of> 6,350 gpm with 122 psid between the reactor and DW (128 psid between the RPV and SP) per TS 3.5.1 and 3.5.2 can be reduced to an analytical flow rate of 5,625 gpm with 122 psid between the reactor and DW (128 psid between RPV and SP), bounding the proposed TS value of > 6,200 gpm with a 122 psid between the reactor and DW (128 psid between the RPV and SP).

Evaluations for non-LOCA analyses confirms that the RHR/LPCI and LPCS proposed TS values of 7,200 gpm and 6,200 gpm, supported by analytical values of 6,713 gpm and 5,625 gpm respectively, does not cause existing results to be invalid or they have an insignificant effect based on assessment results.

Additionally, the following plant changes or flexibility options are not impacted and the conclusions from this report are applicable to the flexibility options as well as nominal operating conditions:

"PRNM and ARTS/MELLLA

  • Elimination of Selected Response Time Requirements (NEDO-32291)

"Standby Liquid Control System (Boron Concentration increase)

"MSIV Closure on Level 1 The GE RHR System Design Specification for Columbia remains applicable as contributing to the Columbia RHR system TS value calculations, assuming no other changes were made in the associated plant configuration.

The adequacy of the ECCS design, as described in FSAR Section 6.3.3 (including acceptance criteria in FSAR Section 6.3.3.2), is unaffected by lowering the LPCS and RHR/LPCI flow rates.

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NEDO-33813 Revision 2 Non-Proprietary Information - Class I (Public) 6.0 References

1. "Columbia Generating Station Final Safety Analysis Report," Chapter 6, Engineered Safety Features, Amendment 59, December 2007.
2. GEH Safety Communication SC 06-01, "Worst Single Failure for Suppression Pool Temperature Analysis," January 19, 2006.
3. General Electric Company, "The GE General Electric Pressure Suppression Containment System Analytical Model," NEDO- 10320, March 1971.
4. General Electric Company, "The General Electric Mark III Pressure Suppression Containment System Analytical Model," NEDO-20533, June 1974.
5. Letter from Ashok Thadani (NRC), to Gary L. Sozzi (GE), "Use of SHEX Computer Program and ANSI/ANS 5.1-1979 Decay Heat Source Term for Containment Long-Term Pressure and Temperature Analysis," July 13, 1993.
6. GE Nuclear Energy SIL Number 636, "Additional Terms Included in Reactor Decay Heat Calculations," Revision 1, June 6, 2001.
7. Moody, F. J., "Maximum Flow Rate of A Single Component Two-Phase Mixture," Journal of Heat Transfer, Transaction of the ASME, Paper No. 64-H6-35.
8. GE Hitachi Nuclear Energy, "Columbia Generating Station GE14 ECCS-LOCA Evaluation," 0000-0090-6853-RO, Revision 0, February 2009.
9. "Columbia Generating Station Final Safety Analysis Report," Chapter 15, Accident Analyses, Amendment 59, December 2007.
10. GE Hitachi Nuclear Energy, "Energy Northwest Columbia Generating Station APRM/RBM/Technical Specifications / Maximum Extended Load Line Limit Analysis (ARTS/MELLLA)," NEDC-33507P, Revision 1, January 2012.
11. GE Nuclear Energy, "WNP-2 Power Uprate Project NSSS Engineering Report,"

GE-NE-208-17-0993, Revision 1, December 1994.

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